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|Alcator C-Mod Opens the 55th APS-DPP Conference|
|The most prominent components visible in the panorama above are the J-Port Field Aligned ICRF antenna on the left, and the two two-strap ICRF antennas on the right (D- and E-Port antennas). The copper plated Inconel straps, seen behind Faraday screens, carry the 50-80 MHz RF currents. These antennas have successfully coupled up to 6 MW of RF power into C-Mod plasmas. The J-Port antenna straps are oriented in 3D so that RF currents flow perpendicular to the magnetic field. This rotation resulted in dramatically reduced impurity generation relative to the D- and E-Port antennas. Toward the center or the photo is the lower hybrid launcher, which has delivered up to 1 MW of power at 4.6 GHz to the plasma and has produced discharges with the plasma current driven fully non-inductively. The QCM antenna on the left is used to measure and drive turbulent modes. The panorama was composed of 276 photos taken with a n automated camera system, which were exposure blended and stitched into a composite image.|
USBPO Mission Statement: Advance the scientific understanding of burning plasmas and ensure the greatest benefit from a burning plasma experiment by coordinating relevant U.S. fusion research with broad community participation.
Director's Corner C.M. Greenfield USBPO Topical Group Highlights Modeling 100% Non-Inductive Scenarios for NSTX-Upgrade
S.P. Gerhardt, C. Kessel, F. Poli, and R. Raman
ITPA Update Announcements Schedule of Burning Plasma Events Contact and Contribution Information
by C.M. Greenfield
Taking a step back to consider...why we are doing this?
Over the coming months, you’ll be hearing a lot about the politics surrounding ITER. I’d like us to take a small step backwards and think about why we decided a decade ago that this was the path our field of science needs to take.
The quest for fusion as a power source has driven the development of an entirely new branch of physics concerned with the behavior of the plasma state that makes up most of the mass of the universe. Experiments up to and including those currently operating have demonstrated sustained and significant progress toward the goal of fusion energy.
The next frontier for fusion science is the study of the burning plasma regime, where the fusion process itself provides the dominant heat source for sustaining the plasma heat content (i.e., self-heating). The “burning plasma” regime for deuterium-tritium (D-T) plasmas is entered when Q≥5, where Q is the ratio of output fusion power to input heating power – a condition that implies at least 50% of the heating is provided by the ultra-energetic alpha particles created from D-T fusion reactions. It is anticipated that many physics features of fusion plasmas in such self-heated conditions will be very different from those encountered on present day experiments in which the plasma is heated primarily by external means. Achieving and sustaining burning plasmas will have the dual benefits of proving the technical basis for fusion energy production, as well as opening up access to further advancement of plasma science.
ITER will provide the first opportunity to study burning plasmas in the laboratory, enabling advances in plasma and materials sciences. Some of the scientific questions that we will address in ITER include:
- What is the behavior of a large population of energetic alpha particles in the bulk DT plasma?
- The products of the DT fusion reaction are a 14.1 MeV neutron and a 3.5 MeV alpha particle (helium nucleus) that will remain in and heat the plasma.
- How will a self-organized plasma develop, in which most of the heating power is provided by the fusion reactions themselves?
- With dominant self-heating from fusion reactions, a burning plasma determines its own profiles.
- The critical elements in the areas of transport, stability, boundary physics, energetic particles, heating, etc., will be strongly coupled nonlinearly in a burning plasma due to the fusion self-heating.
- New phenomena arise from full nonlinear interplay of alpha particle heating with transport, stability, and current/pressure control, as well as their compatibility with a divertor and plasma-facing materials in steady-state conditions.
- Multi-physics, multi-scale integrated behavior, which cannot always be anticipated from tests and simulations of separate effects, will occur.
- What control schemes and actuators will be effective for controlling that self-organized state and allowing optimization of that state?
- We will need “smart” control, where minimal power can be used to tailor the profiles.
- Example: Control of the current profile with electron cyclotron current drive, which could in turn modify the temperature and density profiles through nonlinear couplings.
- Burn control will be a new challenge: In this self-organized state, we will need to maintain conditions for constant fusion power output.
- How will materials behave in the burning plasma environment?
- Energy density and energy throughput are very high in a burning plasma.
- Surface heat fluxes are higher than in present experiments, and transients must be well controlled.
- Dust production and tritium inventory must be kept to a minimum.
- Neutron flux to both first wall and blanket materials will present new challenges.
- How will we deal with the measurement challenges in this new, harsh, environment?
- Many present-day plasma diagnostic techniques will not be usable in the burning plasma environment due to high neutron/gamma/plasma heat fluxes and limited access.
The behavior of burning plasmas presents a grand challenge on its own scientific merit and is a critical next step on the way to deployment of fusion as an energy source. The knowledge gained will be needed for, and relevant to, any future magnetic fusion reactor.
ITER will provide a unique and essential laboratory to address these grand challenges.
ITER Science and Technology Advisory Committee Meeting
The Science and Technology Advisory Committee (STAC) of the ITER Council held its fifteenth meeting October 15-17. This was an unusual meeting for the US participants, as most of us were unable to travel to France due to the government shutdown. Nevertheless, through the magic of videoconferencing, Earl Marmar, Juergen Rapp, Jim Van Dam, and I were able to participate in the entire meeting from our homes and offices. Rob Goldston did participate in person, since he had already been in Europe before the meeting.
The Council at its June 2013 meeting (IC-12) had formulated a set of six charges to STAC-15. Here is a condensed version:
- Assess the technical aspects of the Level-0 Reference Schedule
- Assess the progress in updating the ITER Research Plan with expert community input
- Assess progress in the analysis of carbon and tungsten divertor options and provide recommendations for the first ITER divertor choice
- Assess progress in the design and prototype manufacturing of the in-vessel coils as well as in the plans for the associated Final Design Review
- Assess progress in the characterization of disruptions and runaways and their mitigation
- Assess status reports on the progress made on key open technical issues (includes ECH, ICH, neutral beams, and neutronics)
As was the case at STAC-14, a nice summary of the STAC-15 meeting was published in the ITER eNews (http://www.iter.org/newsline/286/1739). There was also a story in Science (http://news.sciencemag.org/europe/2013/10/hot-news-fusion-researchers-recommend-iter-design-tweaks).
A major topic of discussion was the schedule, with several critical path items being behind schedule. The STAC recommended that the ITER Organization, in close collaboration with the Domestic Agencies, make an all-out effort to stem the slippages in the schedule and develop credible recovery plans.
The ITER Research Plan (IRP) was also discussed. An international workshop (http://www.iter.org/newsline/283/1719) had been convened in September to reassess the plan in light of changing anticipated availability for many components (US participants were Don Hillis, Stan Kaye, Mickey Wade, and Steve Wolfe). This workshop was a chance for these experts to weigh in on the physics requirements and sequencing of research for improved consistency with the schedule. The STAC received a report from this workshop.
We also continued our discussion from STAC-14 of two major systems that are under consideration for inclusion in the ITER baseline design: Tungsten divertor targets, and in-vessel coils for ELM control (technical reports on each will be given at the ITER Town Meeting during APS; see below). In both cases, the STAC reiterated the endorsements given at STAC-14 and urged that R&D should continue on some outstanding issues.
Another major system that is rising to the forefront of the discussion is the disruption mitigation system (DMS). The STAC urged that adequate resources be allocated (this is really a request to the world’s tokamaks) to research needed to specify the DMS. There will be a lot more discussion of the DMS as we move toward a final design review of that system in Spring 2017.
Other topics of discussion included heating systems and neutronics and shielding.
The output of STAC-15 is a report containing our recommendations that will be taken up by the ITER Council at its upcoming meeting on November 20-21. They will also consider recommendations from the Management Advisory Committee’s (MAC) meeting earlier this week.
Plans for APS-DPP conference
For the sixth straight year, the US Burning Plasma Organization has organized a contributed oral session on Research in Support of ITER on Wednesday afternoon of the 55th Annual Meeting of the Division of Plasma Physics, which will take place in Denver, Colorado, on November 11-15. The agenda for this session can be found here:
The USBPO is also organizing a Town Meeting on ITER, scheduled for Thursday evening (November 14) in the Sheraton Denver Downtown Hotel. We will have a compelling program, focusing on two ITER design decisions to be formalized late this year.
The agenda include the following on development of ITER systems:
Richard Pitts (ITER Organization):
Physics basis and design of the ITER full tungsten divertor
Edward Daly (Thomas Jefferson National Accelerator Facility):
ITER In-Vessel Coils - Design and Status
David Rasmussen (for Ned Sauthoff, US ITER Project Office):
US ITER project status
Although not a USBPO activity, I would also like to call your attention to the annual meeting of the University Fusion Association, held during the APS-DPP conference on Monday evening. Everyone is welcome to attend. Ed Synakowski (DOE) will give the keynote presentation, on the state of DOE FES funding and areas of priority in research. This will be followed by two short presentations, the first by Dale Meade (Fusion Innovation Research and Energy) titled “Framework for a Road Map to Magnetic Fusion Energy,” and the second by Michael Mauel (Columbia U and USBPO Council Chair Emeritus) titled “Building Consensus in the Formation of Science Strategy: Reflections on the U.S. Fusion Science Program.”
WebinarsA USBPO webinar is scheduled for Wedenesday, November 6, on “Recent activities of the ITPA Transport and Confinement Topical Group.” The presenters will be George McKee (University of Wisconsin) and Gary Staebler (General Atomics), who are the leaders of the USBPO Transport and Confinement Topical Group. You should have received a message from Amanda Hubbard on how to participate (a link to instructions also appears later in this issue of eNews). We are also working on scheduling additional web seminars to report on the recent spate of ITPA topical group meetings.
Editor Note: The BPO Integrated Scenarios Topical Group works to facilitate U.S. efforts to understand, improve, and predict the behavior of whole-device operation (leaders are Stefan Gerhardt and Chris Holcomb). This month's Research Highlight by S.P. Gerhardt, et al., describes modeling of non-inductive scenarios for the NSTX-U device presently under construction at PPPL. A wide range of auxiliary heating and plasma profile options are modeled, demonstrating the advanced state of theoretical and computational tokamak models available.
Modeling 100% Non-Inductive Scenarios for NSTX-Upgrade
S.P. Gerhardt1, C. Kessel1, F. Poli1, and R. Raman2
1Princeton Plasma Physics Laboratory, Princeton, NJ, USA
2University of Washington, Seattle, WA, USA
The National Spherical Torus Experiment (NSTX) at PPPL is presently undergoing a significant upgrade . The first major component of this upgrade is the fabrication of a new “center stack”, containing the Ohmic heating (OH) solenoid and toroidal field (TF) coil inner legs. This will allow an extension of the toroidal field from BT~0.5 T for ~1 second duration, to BT=1 T for ~6 second duration. Similarly, the new OH coil has a factor of ~3 increase in the available flux, designed to allow IP=2 MA plasmas to be sustained for 5 seconds. The second major component of the upgrade is the addition of a second neutral beam system, containing three neutral beam sources to supplement the three sources on the existing beamline. This upgrade doubles the available power to the plasma, with up to Pinj=18 MW available for ~1 second pulses or 10.2 MW available for 5 second pulses. Equally important, the injection geometry of the new beamline is arranged such that two of the three new sources can provide substantial off-axis neutral beam current drive. The neutral beam current drive from these beam systems, when added to the bootstrap current, is designed to support 100% non-inductive operation at plasma currents up to IP=1.3 MA, as described below. Other components of the NSTX-Upgrade project are described in Ref. .
In order to fully anticipate the operational space of NSTX-U fully non-inductive plasma scenarios, a large number of free-boundary TRANSP calculations of fully relaxed equilibria have been executed . These calculations assume fixed, experimentally derived shapes for the electron density and temperature profiles, and flat Zeff profiles. Quasi-neutrality and the assumption of neoclassical ion thermal transport (using the Chang-Hinton model ), allow the ion density and temperature profiles to be determined. The electron density profile is scaled to achieve a given Greenwald fraction ƒGW = n̄e/(Ip/πa2 , while the electron temperature profile is scaled so that the resulting thermal confinement matches either the ITER-98(y,2) or spherical torus (ST) specific sca ling expressions. The use of ITER-scaling is justified by the observation that discharges with lithium plasma facing component (PFC) conditioning [5,6] have tended to follow this scaling [7,8]. The ST-specific scaling expression , with a much stronger BT dependence, is based on data with boronized PFCs in NSTX, and results from MAST show similar parametric dependencies . Understanding which scaling expression holds for lower collisionality ST plasmas is a key goal of the NSTX-U research program.
A set of example calculations is shown in Fig. 1 for 100% non-inductive scenarios at BT=1.0 T, an injected power of Pinj=12.6 MW, and Greenwald fraction of 0.72. This beam power is achieved by using each of the six neutral beam sources at 2.1 MW injected power (Vinj=90 kV), for which the allowed beam duration is 3.0 seconds. The non-inductive current level has been found using both thermal confinement scaling expressions, and for both broad and narrow profile shapes. For the broad profiles and ITER-98(y,2 ) scaling, the non-inductive current level is found to be IP~975 kA, at βN=4.3 and qmin=1.5. More peaked profiles result in a reduction of the bootstrap current, such that the non-inductive current level is reduced to IP=875 kA. Note that the broad profile Te and ne shapes were taken from a discharge designed to operate NSTX at higher-aspect ratio in order to prototype discharges that would be run on NSTX-U , and are thus justified for use in extrapolation to NSTX-U.
Figure 1: Impact of the confinement level and profile shapes on the 100% non-inductive operating scenarios of NSTX-Upgrade. Shown are a) the electron temperature profile, b) the electron density profile, and c) the safety factor profile. The insets show the confinement scaling expressions and information about the configurations.
These results are to be contrasted with the calculations assuming the ST-specific confinement scaling expression. The stronger BT scaling in this expression results in significantly better confinement, with the result that the non-inductive current levels are significantly higher, in the vicinity of 1300 kA for both broad and narrow profiles. For fixed Greenwald fraction, this results in a significantly higher total density, and significantly higher total stored energy. Furthermore, the values of qmin and βN are substantially higher, and these scenarios would likely challenge the global stability of the configurat ion if they are achieved.
The relative roles of density and global confinement are further examined in Fig. 2. Here, contours of non-inductive current fraction and minimum safety factor are shown vs. H98(y,2) (H98(y,2) is the fraction by which the thermal confinement exceeds the level predicted by the ITER-98(y,2) scaling expression) and Greenwald fraction. In this case, the plasma current is fixed at IP=1 MA, with a toroidal field of BT=1.0 T and an injected power of 12.6 MW, and the broad thermal profiles from Fig. 1 are used. For the range of densities considered, the configuration is 100% non-inductive when H98(y,2)~1.05; this apparent density independence of the non-inductive current drive is due to the increase in bootstrap current at higher density, even as the beam current drive is reduced. However, there is a reduction in the required confinement at lower density, as the efficiency of neutral beam current drive increases.While the non-inductive fraction is roughly independent of density in this parameter range, the underlying current profile is not. As the density is reduced, the on-axis neutral beam current drive becomes quite large, resulting in a reduction of the central safety factor. In particular, qmin drops beneath unity for fGW<0.6-0.7 for these confinement multipliers and engineering parameters. These low qmin
Figure 2: Impact of the line-average density and global confinement on the a) non-inductive current fraction and b) minimum safety factor.
|Figure 3: Example modification of the safety factor profile for IP=800 kA, BT=1.0 T, H98(y,2)=1, fGW=0.7 scenarios, using different combinations of four neutral beams. Frame a) shows the profile of neutral beam drive currents, while frame b) shows the safety factor profile. The caption indicates the beam tangency radius, the minimum safety factor and the non-inductive fraction.|
values will result in the onset of strong core MHD , and this may result in an effective lower density limit for fully relaxed scenarios with these parameters.
The above calculations illustrate examples where all six neutral beam sources are used. By properly choosing a subset of the available sources, however, it is possible to control aspects of the current profile . This is illustrated in Fig. 3 for scenarios with four total neutral beam sources. When using the combination of sources with the highest current drive efficiency (green), this configuration is 100% non-inductive at IP=800 kA, with qmin=1.51. When the source with the most off-axis current drive is replaced by one with more central current drive (black), the non-inductive fraction is barely changed, but the central safety factor drops to 1.1. At the other extreme, a combination of sources that minimizes the central current drive to the largest extent possible (red) results in qmin~2.4, albeit with a reduction of the non-inductive current fraction to ~87%. Hence, given a realtime assessment of the current profile and an operating point such as in Fig. 3, realtime control of the central or minimum safety factor should be possible.
The above studies have focused on the fully relaxed non-inductive state of plasmas that likely had an inductive formation process. However, modeling exercises are also examining non-inductive formation and IP ramp-up techniques. The initial plasma is envisioned to be formed by transient Coaxial Helicity Injection (CHI) . This technique has demonstrated the formation of up to ~300 kA plasmas in NSTX, and the larger field and improved magnetic geometry in NSTX-U result in projections of even higher closed-flux currents [1,12]. However, in order to couple these plasma to further heating and current drive methods, it will likely be necessary to increase the electron temperature and slow their resistive decay. It is envisioned that a 1 MW, 28 GHz gyrotron system will be brought into operation once NSTX-U is running, in order to provide electron cyclotron heating of these CHI plasmas. TSC simulations indicate that even with L-mode scaling, it should be possible to bring the electron temperature to the Te=200-400 eV range within ~20 ms. High-Harmonic Fast Wave (HHFW) and neutral beam current drive would then be used to ramp the plasma current to full value. Note that very high non-inductive fractions have previously been observed in 300 kA HHFW heated discharges , and that fast ions from the 2nd neutral beam are very well confined, even for quite low values of the plasma current .
Figure 4: Components of the total current (top), and the power levels of the various heating and current drive actuators available in NSTX-U (bottom), for a TSC simulation of 100% non-inductive current ramp-up using the NSTX vessel geometry. The outer poloidal field coils and outer vessel geometry are the same for both machines.
This process is demonstrated in the TSC simulations in Figs. 4 and 5. These simulations use the Coppi-Tang transport model, and utilize the full complement of available and planned heating and current drive actuators. Considering the plasma current components in the top frame, the initial plasma is provided by CHI, in this case as a 400 kA target at t=17 ms . This plasma is immediately heated by 500 kW absorbed ECH to slow its decay. The fast wave system is also turned on, ramped to a power of 4 MW at 200 ms and held constant at this level until t=325 ms. These actuators lead to an increase in the current from bootstrap and HHFW driven currents. Starting from 300ms, the neutral beam power is ramped up, and as the plasma continues to become hotter and denser, the neutral beam and bootstrap currents become dominant. The fast wave system is simultaneously turned off, to avoid acceleration of NB fast ions . The plasma current achieves the IP=1 MA level after ~4.5 seconds.
Additional parameters related to this simulation are shown if Fig. 5. The top frame shows the Greenwald density fraction and internal inductance; the Greenwald fraction is maintained at fGW=0.5, implying a ramping absolute density as the current increases. The confinement parameters are shown in frame b). Typical confinement times in this simulation are ~40 ms, corresponding to H98(y,2)=1. Here we have assumed the onset of an H-mode at 325 ms with the injection of the HHFW power, as observed in NSTX experiments. Hence, exceptionally good confinement is not a requirement for non-inductive ramp-up. Finally, the electron and ion temperature predicted for this simulation are 1.7 and 2.8 keV respectively.
The NSTX-U program  has additional goals beyond the non-inductive ramp-up and high-beta sustainment research described here. A key goal will be to expand confinement studies to higher field and current, with associated lower collisionality, in order to improve the extrapolation of transport to next-step STs; the discussion around Fig. 1 shows that the ST-specific confinement scaling expression has favorable dpendencies, but its range of validity and underlying physics must be better understood. Additional research will focus
|Figure 5: Additional quantities related to the simulation in Fig. 4. Shown are a) the Greenwald density fraction and normalized internal inductance, b) the energy confinement time and H98(y,2) multiplier, and c) the electron and ion temperatures.|
development of strategies for divertor heat flux control and the core impurity content, and the assessment of high-Z metals and lithium surfaces as plasma facing components. The first plasma in NSTX-U is presently scheduled for late 2014, with research following shortly afterwards.
 J.E. Menard, et al., Nuclear Fusion 52, 083015 (2012)
 S.P. Gerhardt, et al., Nuclear Fusion 52, 083020 (2012)
 C.S. Chang and F.J. Hinton, Phys. Plasmas 25, 1493 (1982)
 M. Greenwald, Plasma Phys. Control Fusion 44, R27 (2002)
 H. Kugel, et al., Phys. Plasmas 15, 056118 (2008)
 R. Maingi, et al., Nuclear Fusion 52, 083001 (2012)
 S.P. Gerhardt, et al., Nuclear Fusion 51, 073031 (2011)
 S.M. Make, et al, Nuclear Fusion 53, 063005 (2013)
 S.M. Kaye, et al., Nuclear Fusion 46, 848 (2006)
 M. Valovic, et al., Nuclear Fusion 49, 075016 (2009)
 S.P. Gerhardt, et al., Nuclear Fusion 53, 043020 (2013)
 R. Raman, et al., Phys. Plasmas 18, 092504 (2011)
 G. Taylor et al., Phys. Plasmas 19, (2012) 042501
 R. Raman, D. Mueller, S.C. Jardin, T.R. Jarboe, et al., Nuclear Fusion 53, 073017 (2013)
 D. Liu, et al., Plasma Phys. Control Fusion 52, 025006 (2010)
 NSTX-U Five Year Research Plan: http://nstx-u.pppl.gov/five-year-plan/five-year-plan-2014-18
|4th Meeting, ITER Site, France, December 9 - 11, 2013|
|Diagnostics Topical Group|
25th Meeting, ITER Site, France, October 16 - 18, 2013
|Energetic Particle Physics Topical Group|
The 11th ITPA meeting was held in Beijing, China immediately following the IAEA technical meeting on energetic particles. S.D. Pinches opened the meeting with a talk on EP physics development in ITER and the priority needs to make sure the risk to make big investment in ITER project will be minimal. He pointed out the first delivery of the dummy conductor, produced in China, took place and that it underwent impressive mock up schematic tests. Several diagnostics are being prepared by the ITER contributing parties. Getting the most attention are the ones aimed at studies of the major physics topic of the confinement of EPs in the presence of Alfvénic activity. The diagnostic technique considered for this is an extensive Mirnov coil array capable of measuring not only toroidal and poloidal mode numbers, but also the polarization of the oscillations. This system is on track to be available at the start of ITER operations. It was noted that use of an ICRH antenna to detect the various types EP modes (including the high frequency responsible for ion cyclotron emission) is also planned.
|Integrated Operation Scenarios Topical Group|
This October, the ITPA Integrated Operations Scenarios (IOS) meeting focused on advanced scenario experiments, simulations of the ITER baseline scenario (IBS) and reports from Joint Experiments. The meeting also had three parallel joint sessions with the Pedestal and with the Transport and Confinement groups on (1) pedestal projections and impact on integrated modeling, (2) including energy transport in integrated modeling, and (3) integrated modeling of fueling and impurities. Common ground in these sessions was the need of the IOS to get advice on which models to use in our integrated simulations. [The full summary by F. Poli is available at the forum post linked above. -Ed.]
|MHD, Disruptions & Control Topical Group|
22nd Meeting, Hefei, China, October 8 - 11, 2013
|Pedestal & Edge Physics Topical Group|
|25th Meeting, Kyushu University, Japan, October 7 - 9, 2013|
|Scrape-Off-Layer &a mp; Divertor Topical Group|
|18th Meeting, Hefei, China, March 19 - 22, 2013|
|Transport & Confinement Topical Group|
|11th Meeting, Fukuoka, Japan, October 7 - 9, 2013 |
Areas to be covered include impurity and particle transport; validation of gyrofluid transport models; momentum transport; transport in the L-mode edge, particularly during the current rise phase of ITER; L-H and H-L transitions; profile stiffness; 3D effects; and the long-term effort to provide a fully validated model of plasma transport for ITER. These areas include topics that have been selected for special reports to the Integrated Operation Scenarios Topical Group .
BPO Web Seminar
The USBPO is pleased to announce the next in its series of web seminars on subjects of interest to the fusion community.
Time: Wednesday, Nov 6th, 1 pm EST, 12:00 CST, 10:00 PST.
Topic: Recent activities of the ITPA Transport and Confinement Topical Group
Presenters: George McKee, Univ. Wisconsin and Gary Staebler, General Atomics,
leaders of the USBPO Confinement and Transport Topical Group
Remote Connection Info Available at https://burningplasma.org/forum/index.php?showtopic=1279
They will review some current issues and highlights in this topical area, following the ITPA meeting held Oct 7-9 in Kyushu, Japan. As usual there will be plenty of time for discussion.
|November 6, BPO Web Seminar: POSTPONED|
Recent activities of the ITPA Transport and Confinement Topical Group
Presented by G.R. McKee and G. Staebler
Remote connection info available at https://burningplasma.org/forum/index.php?showtopic=1279ia
|November 11 - 15, APS DPP Meeting, Denver, United States|
|November 18 - 20, 18th Workshop on MHD Stability Control, Santa Fe, New Mexico, USA|
|December 9 - 11, ITPA: 4th CC/CTP Meeting, ITER|
|December 11, 4th CTP ExCom Meeting, ITER|
|December 16 - 20, IAEA: 2nd DEMO Programme Workshop, Vienna, Austria|
NSTX-U commissioning operations begin
November, First plasma at ITER
First plasma at W7-X
March, Beginning of full DT-operation at ITER
First plasma at JT-60SA
This newsletter provides a monthly update on U.S. Burning Plasma Organization activities. Topical Group Highlight articles are selected by the Leader and Deputy Leader of those groups (burningplasma.org/organization/?article=Topical%20Groups). ITPA Reports are solicited by the Editor based on recently held meetings. Announcements, Upcoming Burning Plasma Events, and all comments may be sent to the Editor. Suggestions for the Image of the Month may be sent to the Editor. The images should be photos, as opposed to data plots, though combined graphics are welcome. The goal is to highlight U.S. fusion resources through interesting visualizations.
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